Text

Proposed Tech Specs 3/4.4.8.1, Pressure/Temp Limits, 3.4.8.3.1, Overpressure Protection Sys-RCS Temp Less than or Equal to 312 F & 3.4.8.3.2, Overpressure Protection Sys-RCS Temp Greater than 312 F
	ML20071F822
Person / Time
Site:	San Onofre
Issue date:	07/06/1994
From:	; SOUTHERN CALIFORNIA EDISON CO.
To:	;
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ML20071F811	List: ML20071F809; ML20071F814; ML20071F822;
References
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ATTACHMENT A EX1 STING TECHNICAL SPECiflCATIONS AND BASES l UNIT 2 i i

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9407110184 940706 -

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i INDEX LIMITING CON 0! TION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE HOT SHUTD0WN............................................ 3/4 4-3 COLD SHUTDOWN - Loops filled............................ 3/4 4-5 l 2

COLD SHUTDOWN - Loops Not Filled........................ 3/4 4-6 3/4.4.2 SAFETY VALVES - 0PERATING............................... 3/4 4-7 3/4.4.3 PRESSURIZER............................................. 3/4 4-8 3/4.4.4 STEAM GENERATORS........................................ 3/4 4-9 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS............................ 3/4 4-16 OPERATIONAL LEAKAGE.................................. 3/4 4-17 3/4.4.6 CHEMISTRY............................................... 3/4 4-20 3/4.4.7 SPECIFIC ACTIVITY....................................... 3/4 4-23 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR CCOLANT SYSTEM............................... 3/4 4-27 PRESSURIZER - HEATUP/C00LDOWN........................ 3/4 4-31 OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE < 312*F............................ -3/4 4 RCS TEMPERATURE I 312'F............................ 3/4 4-33 3/4.4.9 STRUCTURAL INTEGRITY.................................... 3/4 4-34 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-35 3/4.5 EMERGENCY' CORE COOLING SYSTEMS i

l 3/4.5.1 SAFETY INJECTION TANKS.................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350'F..........................- 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350*F.......................... 3/4 5 -3/4.5.4 REFUELING WATER STORAGE TANK......-.. ................... 3/4 5-8 t

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SAN ON07RE-UNIT 2 V- AMEN 0 MENT M).JO t 2

IN It LIST 08

AS' !S TABLE DAGE 3.3-10 ACCIDENT MCNITORING INSTRUMENTATION. . . . . . . . . . . . . . . . . . . . . . 3/4 3-52 4.3 7 AC010ENT SONITORING INSTRUMENTATION SURVEILLANCE R!;UIREMENTS............................................. 3/4 3-54 3.3-1; FIRE ;ETE;TICH INSTRUMENTS MINIMUM INSTRUMENTS OPERABLE 3/4 3-57 3.3-12 RA0:0 ACTIVE LIQUID EFFLUENT MON!iCRING INSTRUMENTATION --

DELETED 4.3 8 RADICACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION l SURVEILLANCE REQUIREMENTS - DELETED i 3.3-13 RADICACTIVE GASECUS EFFLUENT MONITORING INSTRUMENTATION., 3/4 3-55 4.3 9 RADI0 ACTIVE GASECUS EFFLUENT MONITORING INSTRUMENTATION i

SURVEILLANCE REQUIREMENTS................................ 3/4.3-67 l 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED OURING INSERVICE INSPECTION..................................... 3/4 4-14 4.4 2 STEAM GENERATOR TUBE INSPECTION.......................... 3/4 4 15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES. . . . . . . . . ,

3/4 4-19 3.4 2 REACTOR COO LANT SYSTEM CHEMISTRY. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-21 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE. . . . . . . . 3/4 4-30a ,

4.4-i REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................................. 3/4 4 22 4.4 4 PRIMARY C0OLANT SPECIFIC ACTIVITY SAMPLI AND ANALYSIS PR0 GRAM.................................................. 3/4 4 25

-i 4.4-5 - REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITH0RAWAL SCHEDULE...................................... 3/4 4-28 4.6-1 TENDON SURVEILLANCE...................................... 3/4 6-12 4.6 2 - T E NDON L I FT- 0 F F F 0 R C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-12a 3.6-1 CONTAINMENT ISO LATION VALVES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-20 3.7-1 MAIN STEAM SAFETY VALVES................................. 3/4 7-2' 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP WITH INCPERA8LE MAIN STEAM SAFETY YALVES DURING OPERATION '

WITH BOTH STEAM GENERATORS............................... 3/4 7 SAN ON0FRE-UNIT 2 XIX AMEN 0 MENT NO. 91 n-m,- ~ w,,_,,,m+-,,--,,n,,,-n,- , , . - , . ,-a , . , - - - - - .c,_-- - - , , , ,,m,-, , , , - - , ~ , ., ,- ,e- ,f,,-

INDEX LIST OF FIGURES FIGURES PAGE 3.I-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AND TEMPERATURES AS A FUNCTION OF STORED BORIC ACID CONCENTRATION......... 3/4 1-13 '

3.1-2 CEA INSERTION LIMITS..................................... 3/4 1-24 3.2-1 DNBR MARGIN OPERATING LIMIT BASED ON C0LSS............... 3/4 2-7 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE). . . . . . . . . . . . . . . . . . . . . . . 3/4 2-8 3.3-1 DEGRADED BUS VOLTAGE TRIP SETTING........................ 3/4 3 4.4-1 TUBE WALL THINNING ACCEPTANCE CRITERIA................... 3/4 4-15a 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT........................................... 3/4 4-26 3.4-2 RCS NEATUP PRESSURE / TEMPERATURE LIMITATIONS FOR 4-8 EFPY............................................. 3/4 4-29 3.4-3 RCS C00LDOWN PRESSURE / TEMPERATURE LIMITATIONS FOR 4-8 EFPY................................................. 3/4 4-30 3.7-1 MINIMUM REQUIRED FEE 0 WATER INVENTORY FOR TANK T-121 FOR MAXIMUM POWER ACHIEVED TO DATE............ 3/4 7-6A l 5.1-1 EXCLUSION AREA........................................... 5-2 t

5.1-2 LOW POPULATION Z0NE...................................... 5-3 5.1-3 SITE BOUNDARY FOR GASEOUS EFFLUENTS...................... 5-4 l

5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS....................... 5-5 5.6-1 UNITS 2 & 3 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT l FO R R EG ION I I RAC KS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5- 12 5.6-2 UNIT 1 FUEL

MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION II RACKS...................................... 5-13 5.6-3 FUEL STORAGE PATTERNS FOR REGION II RACKS. . . . . . . . . . . . . . . . 5-14 5.6-4 FUEL STORAGE PATTERNS FOR REGION II RACKS RECONSTITUTION STATI0N................................... 5-15 6.2-1 0FFSITE ORGANIZATION..................................... 6-2 6.2-2 UNIT ORGANIZATION........................................ 6-3 6.2-3 CONTROL ROOM AREA........................................ 6-4a SAN ON0FRE-UNIT 2 XXI AMENDMENT NO 87=

l REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.8.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 and Figure 3.4-3 during heatup, cooldown, criticality, boltup, l and inservice leak and hydrostatic testing with:

a. A maximum heatup of 10*F in any one hour period with RC cold leg temperature less than 112*F. A maximum heatup of 30 F in any one hour period with RC cold leg temperature less than 163*F. A maximum heat-upof60FinangonehourperiodwithRCcoldlegtemperature greater than 163 F. .

b. A maximum cooldown of 10*F in any one hour period with RC cold leg temperatures less than 103*F. A maximum cooldown of 30*F in any one hour period with RC cold leg temperatures less than 145"F. A maximum cooldown of 100*F in any one hour period with RC temperature greater than 145 F.

c. A maximum temperature change of less than or equal to 10*F n any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

d. A minimum temperature of 86*F to tension reactor vessel head bolts. l APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY with-in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,yg and pressure to less than 200*F and 500 psia, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system

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heatup, cooldown, and inservice leak and hydrostatic testing operations.

1 SAN ONOFRE - UNIT 2 3/4 4-27 AMENDMENT NO. 70

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS ,

REACTOR COOLANT SYSTiH SURVEILLANCE REQUIREMENTS (Continued) 4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50 Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. Recalculate the Adjusted Reference Temperature based on the greater of the following:

a. The mean value of shift in reference temperature for plates C-6404-3*, or

b. The predicted shift in reference temperature for weld seams 3-203A or '

3-203B as determined by Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988.

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"The most limiting material in the reactor vessel in accordance with the'new l- Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel l Materials," May 1988, has changed and are plates C-6404-3. Calculative proce-dures provided in the new guide should be used to obtain the mean values of shift in RT NDT f.C-6404-3 plates. Calculations are based on the actual shift in reference temperature as determined by impact testing on the existing plate C-6404-2 surveillance material.

-SAN ONOFRE . UNIT 2 3/4.4-27a AMENDMENT NO. 70 l.

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Figure 3.4-2 RCS HEATUP PRESSURE / TEMPERATURE LIMITATIONS F08 4-8 EFPY SAN _ONOFRE __ UNIT-2 3/4 4-29 AMENDMENT NO. 70

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Figure 3.4-3 RCS C00LDOWN PRESSURE /TEMFERATURE LIMITATIONS FOR 4-8 EFPY SAN ONOFRE - UNIT 2 3/4 4-30 AMENDMENT NO. 70

Table 3.4-3 Low Temperature RCS Overpressure Protection Range Operating Period, EFPY Cold Leg Temperature, 'F During During Heatup Cooldown 4 to 10 1 312 1 287 l

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l SAN ONOFRE - UNIT 2 3/4 4-30a AMEN 0 MENT NO. 70-l l

REACTCD CC0t3NT SYSTEM OVEDPPE55URE PDOTECTION SYSTEMS RCS TEMPEPATURE_< 312'F tiMITING CONDITION FOR OPERATION l 3.4.8.3.1 No more than two high-pressure safety qjection pumps shall be OPERABLE and at least one of the following overpres;ure protection systems shall be OPERABLE:

a. The Shutdown Cooling System Relief Valve (PSV)349) with:

1) A lif t setting of 406 i 10 psig*, and

2) Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377, and 2HV9378 open or, l l

b. The Reactor Coolant System depressurized with an RCS vent of greater I than or equal to 5.6 square inches.

APPLICABILITY: MODE 4 when the temperature of any one RCS cold leg is less than or equal to the enable temperatures specified in Table 3.4-3; MODE 5; and MODE 6 when the head is on the reactor vessel and the RCS is not vented.

ACTION:

to less than 200'F,

a. With the SOCS depressurize andRelief vent theValve RCSinoperable, through a redJce g" eaterT,"than or equal to 5.6 square inch vent within the next 8 hr.urs,

b. With one or both SOCS Relief Valve isolation valves in a single SOCS Relief Valve isolation valve pair (valve pair 2HV9337 and 2HV9339 or valve pair 2HV9377 and 2HV9378) clost.d open the closed valve (s) or '

power-lock open the OPEPABLE SOCS Relief Valve isolation valve pair to less- than 200*F, depressurize and within 24 hours, or reduce vent the RCS through a greater T,, than or equal to 5.6 inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. With more than two high-pressure safety injection pumps OPERABLE, secure the third high-pressure safety injection pump by racking out its motor circuit breaker or locking close its discharge valve within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

For valve temperatures less than or equal to 130*F.

AMENDMENT NO. 44,125 SAN ONOFRE - UNIT 2 3/4 4-32

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE > 312*F LIMITING CONDITION FOR OPERATION .

3.4.8.3.2 At least one of the following overpressure protection systems shall !

be OPERABLE: !

a. The Shutdown Cooling System Relief Valve (PSV9349) with:

1) A lift setting of 406 1 10 psig*, and ,

2) Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 open, or,

b. A minimum of one pressurizer code safety valve with a lift setting of 2500 psia + 1%**. ,

APPLICABILITY: MODE 4 with RCS temperature above that specified in Table 3.4-3.

ACTION:

a. With no safety or relief valve OPERABLE, be in COLD SHUTDOWN and vent the. ,

RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

b. In the event the SDCS Relief Valve or an RCS vent is used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the SDCS Relief Valve code safety valve or RCS vent on the transient and any corrective action necessary to prevent recurrence.

SURVEILLANCE REQUIREMENTS 4.4.8.3.2.1 The SDCS Relief Valve shall be demonstrated OPERABLE by:

a. Verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that the SDCS Relief Valve . .

isolation valves 2HV9337, 2HV9339,-2HV9377 and 2HV9378 are open when the SDCS Relief Valve is being used for overpressure protection.

b. Verifying relief valve setpoint at least once per 30 months when ,

tested pursuant to Specification 4.0.5.-

i 4.4.8.3.'2.2 The pressurizer code safety valve has no additional surveillance requirements other than those required by Specification-4.0.5.

4.4.-8,3.2.3 The RCS vent shall- be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months -

when the vent is'being used for overpressure protection, except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

"For valve temperatures less than or equal to 130*F.

l **The lift. setting pressure shall-correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SAN ONOFRE -' UNIT 2 3/4 4-33 AMENDMENT NO. 70 L

L .- - _- - - - - .

REACTOR COOLANT SYSTEM t

BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The heatup and cooldown limit curves (Figures 3.4-2 and 3.4-3) are composite curves which were prepared by determining the most conservative case, with eithar the inside or outside wall controlling, for any heatup rate of up to 60'F/hr or cooldown rate of up to 100'F/hr. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period, and they include adjustments for possible errors in the pressure and temperature sensing instruments.

The reactor vessel materials have been tested to determine their initial RT the results of these tests are shown in Table B 3/4.4-1. Reactor opera-ti$$T;ndresultantfastneutron(EgreaterthanIMev)irradiationwillcause a

an increase in the RT Therefore, an adjusted reference-temperature, based uponthefluenceandUkp.er and nickel content of the material in question, can be predicted using FSAR Table 5.2-5 and the recommendations of Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials."

The heatup and cooldown limit curves, Figures 3.4-2 and 3.4-3, include predicted adjustments for this shift in RT at the end of the applicable service period, aswellasadjustmentsforpossiUNerrorsinthepressureandtemperaturesensing instruments.

The actual shift in RT of the vessel material will be established periodicallyduringoperatigQTby removing and evaluating, in accordance with ASTM E185-73 and 10 CFR 50 Appendix H, reactor vessel material irradiation sur-veillance specimens installed near-the inside wall of the reactor vessel in the core area. The surveillance specimen withdrawal schedule is shown in Table 4.4-5. Since the neutron spectra at the irradiation samples and vessel.

inside radius are essentially identical, the measured transition shift for a- ,

sample can be applied with confidence to the adjacent section of the reactor vessel taking into account the location of the sample closer to the core than the vessel wall by means of the Lead Factor. The heatup and_cooldown curves must be recalculated when the delta RT determined from the surveillance capsuleisdifferentfromthe_calculatIOdeltaRTNDT. the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown-on Figure 3.4-2 and 3.4-3 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

The maximum RT for all_ reactor coolant system pressure-retaining materials, with the NNception of the reactor pressure vessel, has been deter-mined to be 90'F. The Lowest Service Temperature limit line shown on i

l Figure Addenda 3.4-2and3.4-3isbaseduponthisRTl0IlerandPressureVesselCodesince of 1972) of Section III of the ASME Article NB-233 l requires the Lowest Service Temperature to be RT 100*F ' for piping, pumps andvalves.Belowthistemperature,thesystemp$s+uremustbelimitedtoa maximum of 20% of the system's hydrostatic test pressure of 3125 psia.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pres-surizer is operated within the design-criteria assumed-for the fatigue analysis-performed in accordance with the ASME Code requirements.

SAN ONOFRE-UNIT-2 B 3/4 4-7 AMENDMENT NO.70

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TABLE B 3/4.4-1 u, i E REACTOR VESSEL TOUGHNESS

@ Temperature of Minimum Upper l S$ Drop Charpy V-Notch Shelf Cv energy ,

? Weight @ 30 9 50 for Longitudinal d: ' Piece No. Code No. Material Vessel Location Results - f t - Ib - f t - Ib Direction-ft Ib

'!h 215-01 C-6403-1 A533GRBCL-1 Upper Shell Plate 40 15 35 130 25 133

s. 215-01 C-6403-2 o 20 215-01. C-6403-3 -

-10 20 45 131 +

215 C-6404-1 Intermediate Shell Plate -30 10 50 145 215-03 C-6404-2 -20 20 50 155

.215-03 C-6404-3 "

-20 10 50 ,

131 215-02 .C-6404-4: Lower Shell Plate -10 -5 25 124

215-02 C-6404-5 -20 10 25 134 !

215-02 i C-6404-6. -10 -20 -0 151 ,

oo 230-02 C-6401 A508C1-2 Vessel Flange Forging -10 -70 -35 148 t' -209-02 'C-6402 Closure Head Flange -10 -90 -40 142 Forging i s> - .

d, '205-02 C-6410-1 Inlet Nozzle Forging 20 -40 -35 130 ;

205-02 C-6410-2 " '

0 -20 -5 135 205-02 C-6410-3 "

.""' 0 -15 -15 140 ,

205-02 C-6410-4 "

'O -65 -50 140 ;

205-06 C-6411-1 '"

Outlet Nozzle Forging -100 -30 -10 140 ,!

205-06L C6411-2 " "

0 -35 -10 140 l i

232-01 C-6424' A533GRBCL-1 Botton Head Torus -50 -20 10 122 !

232-02 'C-6425 "

Bottom Head Dome -50 -30 -20 136 j 205-03 C-6428-1 A508CL-1 '

Inlet Nozzle Forging S/E -30 -70 -50 174 ;

-205 C-6428-2 " "

-30 -70 -50 . 174 i

'205 .C-6428-3 " "

-30 -70 -50 174 205 C-6428-4 -30 -70 -50 174 .

.. 205-01 'C-6429-1 "

Outlet Nozzle Ext. -30 -40 -25 229 Forging 205 C-6429-1 -30 -40 -25 229 231-02 C-6430-1 A533GRBCL-1 Closure Head Peels +10 20 55 118

- 231-02 C-6431-1 E"- -20 10 50 100 231-02 C-6432-l'

" T" -10 -15 45 115 231-02 C-6432 "

. Closure llead Dome -10 -15 45 115 G _. ._ ,. - - - - , _. . . . _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _

a 4-a%~u+u- -J.6+maA -.-- h- -- ' - ~ - - 4L M'8 - --NO a- * *-* a--h-^ --- * - +=c--""-- # "*

ATTACHMENT B PROPOSED TECilNICAL SPECIFICATIONS AND BASES UNIT 2

c-JNDEX l.IMITING CONDITION 00R OPERAT*0N ANit Ab;j WE kFjgUIREMENTS SECTIDH PAGE HOT SHdTD0WN.......................................... 3/4 4-3 COLD SHUTDOWN - LOOPS FILLE 0.......................... 3/4 4-5 COLD SHUTDOWN - LOOPS NOT FILLED...................... 3/4 4-6 3/4.4.2 SAFETY VALVES - 0PERATING............................. 3/4 4-7 1 3/4.4.3 PRESSURIZER........................................... 3/4 4-8 3/4.4.4 STEAM GENERATORS...................................... 3/4 4-9 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS..................... 3/4 4-16 OPERATIONAL LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-17 ;

3/4.4.6 CHEMISTRY............................................. 3/4 4-20 i

3/4.4.7 SPECIFIC ACTIVITY..................................... 3/4 4-23 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM...................... . 3/4 4-27 PRESSURIZER - HEATUP/C00LDOWN................. 3/4 4-31 OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s 342 256of............. 3/4 4-32 RCS TEMPERATURE > 343 256aF............. 3/4 4-33 52'/#I 3/4.4.9 STRUCTURAL INTEGRITY.................................. 3/4 4-34 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM....................... 3/4 4-35 l 3/4,5 EMERGENCY CORE COOLING SYSTEMS j 3/4.5.1 SAFETY INJECT ION TANKS . . . : . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,y a 350aF......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,y < 350af......................... 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK.......................... 3/4 5-8 I

SAN ONOFRE-UNIT 2 V AMENDMENT NO.

INDEX LIST OF TABtFS TABLE PAGE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION........................ 3/4 3-52 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................... 3/4 3-54 3.3-11 FIRE DETECTION INSTRUMENTS-MINIMUH-INSTRUMENFS-ORERABEE. . . . 3/4 3-57 M -12 R AD10AC44VE-L4QU10-EFRUEN T-MONI40 RING-4NSTRUME44 TAT 40N-

--DELETED-4.3 8 9 AD10AC44 VE-L4 QUID-E FRUENF-MON 140RI-NG-lMSTRUMERTAT40N

--SURVE4 RANGE-REQUIREMENTS DELETED 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION.... 3/4 3-65 [64p?l 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................................. 3/4 3-67 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECT 10N....................................... 3/4 4-14 4.4-2 STEAM GENERATOR TUBE INSPECT 10N............................ 3/4 4-15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES........... 3/4 4-19 3.4-2 REACTOR _ COOLANT SYSTEM CHEMISTRY........................... 3/4 4-21 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE..........

3/44-30adl 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................................... 3/4 4-22 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................... 3/4 4-25 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITHDRAWAL SCHEDULE........................................ 3/4 4-28 4.6-1 TENDON StiRVEILLANCE........................................ 3/4 6-12 4.6-2 TENDON LIFT-OFF FORCE....... .............................. 3/4 6-12a 3.6-1 CONTAINMENT ISOLATION VALVES............................... 3/4 6-20 3.7-1 MAIN STEAM SAFETY VALVES................................... 3/4 7-2 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS................................. 3/4 7-3 SAN ON0FRE-UNIT 2 XIX AMENDMENT NO.

INDEX l tIST OF FIGURES FIGURE PAGE 3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AND TEMPERATURES AS A FUNCTION OF STORED BORIC ACID CONCENTRATION........... 3/4 1-13 3.1-2 CEA INSERTION LIMITS VS FRACTION OF ALLOWABLE THERMAL P0WER................ ..................................... 3/4 1-24 3.2-1 DNBR MARGIN OPERATING LIMIT BASED ON C0LSS................. 3/4 2-7 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE)......................... 3/4 2-8 3.3-1 DEGRADED BUS VOLTAGE TRIP SETTING.................... ..... 3/4 3-40 4.4-1 TUBE WALL THINNING ACCEPTANCE CRITERIA.... ............... 3/4 4-15a 3.4-1 DOSE EQUIVALENT l-131 PRIMARY COOLANT SPECIFIC lgyl ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY >l.0 pCi/ GRAM DOSE EQUIVALENT l-131................. ........ ................ 3/4 4-26 3.4-2 SONGS 2-HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS I FOR-4-8 UNT il 20 EF PY- NORMAL OPERAT10N. . . . . . . . . . . . . . . . . . . . .

3/44-29l6Tl 3.4-3-4 SONGS'2 000LDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS I FOR 4 8 UNTIL 20 EFPY-NORMAL OPERAT10N..................... 3/4 4-30 g l 3.4-5~ SONGS 2 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE 1 l COOLDOWN RATES-(UNTIL 20 EFPY)-NORMAL-OPERATION.............. 3/44-30al%pl 3.4-6 SONGS 2'C00LDOWNRCS PRESSURE / TEMPERATURE LIMITATIONS l UNTIL-20 EFPY-REMOTE SHUTDOWN OPERATION;................... :3/44-30blgy 3.4-7 SONGS:2'RCS PRESSURE / TEMPERATURE 1llMITS MAXIMUM;' ALLOWABLE I C00LDOWN RATES (UNTIL 20 EFPY)-REMOTE SHUTDOWN OPERATION...

3/44-30cl@fI 3.7-1 MINIMUM REQUIRED FEEDWATER INVENTORY FOR TANK T-121 FOR MAXIMUM POWER ACHIEVED TO DATE........ ..... 3/4 7-6A 5.1-1 EXCLUSION AREA............................................. 5-2 5.1-2 LOW POPULATION ZONE...... ................................. 5-3 5.1-3 SITE BOUNDARY FOR GASEOUS EFFLUENTS........................ 5-4 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS........... ...... .. ... 5-5 SAN ON0FRE-UNIT 2 XXI AMENDMENT NO.

JNDEX LIST OF FIGURES FIGURE' .ELG1 5.6-1 UNITS 2 & 3 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION 11 RACKS.. ............. ............. ......... 5-12 l9fI 5.6-2 UNIT 1 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION 11 RACKS........................................ 5-13 5.6-3 FUEL STORAGE PATTERNS FOR REGION 11 RACKS.................. 5-14 5.6-4 FUEL STORAGE PATTERNS FOR REGION 11 RACKS RECONSTITUTION STAT 10N..................................... 5-15 6.2-1 0FFSITE ORGANIZAT10N....................................... 6-2 6.2-2 UNIT ORGANIZATION............ . .... .. ................... 6-3 6.3-3 CONTROL ROOM AREA ... ...... .. . .. . ...... ... .... 6-4a SAN ON0FRE-UNIT 2 XXII AMENDMENT N0.

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTFM llMITING CONDITION FOR OPERATION 3,4.8.1 With the reactor vessel head bolts tensioned',-T-the Reactor l Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, and Figur+

3.4-24, 3.4-5, 3.4-6,- and 3.4-7 during heatup, cooldown, criticality, bokupr and inservice leak and hydrostatic testing with:

a. A-madmtmsheatep-of-10 of in any one hour-per4ed-w-i-14rRC-r444eg t+mperature l e s s t han4-12 o f . f ma*4 mum-heatup-of-30cF-4n-any one hour per4 ed-wi-t h-RC-cM44e9--tempera t u r+4 ess-t-ha n46PFT A maximum heatup of 60of in any one 1-hour period with RCS cold leg temperature greater than 163-or equal to 86oF. l5uf i

b. A-maximum-conidown-of4GoF :n arpne-hour-per4od-wRh-RG-rdd4eg temperatur-es4ess4han403ah A maximum cooldown of-40nf as specified by Figure 3.4-5 in any one 1-hour period with RCS cold leg temperatures less than'or equal to 446 160af. A maximum cooldown of 100af in any one 1-hour period with RCS cold leg temperature greater than 9 l 446 160aF.

c. A maximum temperature change of less-than-or-equal-te loof in any l one 1-hour period during inservice hydrostatic and leak testing i operations above the heatup and cooldown limit curves,

d. A minimum temperature of 86af to tension reactor vessel head bolts.

With the reactor vessel head bolts.detensioned, the Reactor Coolant System (except the pressurizer) temperature shall be limited to a maximum heatup or cooldown of 600F in any 1-hour period.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY with-in the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T and pressure to less than 200of and 500 psia, respectively, within the fEflowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With the reactor vessel head bolts detensioned,'RCS cold leg temperature may !

be less than 86*F.

SAN ONOFRE-UNIT 2 3/4 4-27 AMENDMENT NO.

1 1

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) l 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50 Appendix H in accordance with the schedule in Table 4.4-5. The results of these examinations shall be used to update Figures 3.4-2 and-3v4-313:4E41through13.447. Recalculate the Adjusted Reference :

Temperature based en the'grdatur^cf tiic ~fol'cwing:lin! accordance with Regulatory GuideL 1.99, Revision l:2G" Radiation: Embrittlement: of; Reactor; Vessel Materials,"

May 1988,

a. The mean,value of slMt-in referenee-temperature for plate C - 5403 3-r-or

b. The predic4ed shif t in reference temperature for ;;cid scam: 3 203A+r--

3 203B-+s determined by Regulatory Guide 1.99, Revision 2, " Radiation Embr4t-t4cment of Reactor Ve;;cl Materiah," May 1988.

'The mos t l imi t ing raterial - in th^ -^'"^- "^~ ^' " ' ^~'--^ "h the new Regu htovy-G M de-l.09, Revitica 2, " Radiation Embrittlement of Reacter Vessel Materiah," May 1988, has changed and are plate C 5401 3. Calculative prece dures-prev 44ed-in the nc; guid^ should be used to obtain the =can value; cf s4Fift in RTm -of C 5101 3 plates. Calculation are based en the aetea4-shift -

in-reference temperature as determined-by-impact test 4ng--on the exhting plate C 5404 2 surveillanee-eaterial.

SAN ON0FRE-UNIT 2 3/4 4-27a AMENDMENT N0.

3500 ,

,,,,,,,,,,,,,i,,,.i..,,,,,,,i,,,,i,,,,

LOWEST SERVICE

INSERVICE TESTS # HEATUP TEMP = 209*F !

i i

3Cir)O - - -

i i

Acceptable operating region - to the _

right of the inservice tests curve .

(Apphcable in modes other than i Modes 1 and 2) i-2500 - e~ ++ ~-

y # Acceptable operating region - to the .j . , , , -

g right of the heatup curve in all rnodes. ;

a In addition, in Modes 1 and 2 the ;

g" operating region is to the right of the -l' c core critical curve. I

@u) 2000 - >+ - +- -+

l- - ---

W !'

c -;-

n. .

w .4 . 2. .. . . . . .

gbf l

$ 1500 -

l a

u) * ' '

U) l

. 4 . . . . . 4.. .; . . . .9-.;.4 1 .+ ..7.. .

L]1 '

CORE a

o 3000 CRITICAL ._

w

. so... ... 4 . ,.

. v.. . ,

9 i o _

E

- J

$00 --- -#- ~ < - * ~'"

MINIMUM BOLTUP TEMP = B6*F A

i .iI . iliiiiIiiiiliieilii i liiiilii,,

0 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE (?F)-Tc l FIGURE 3.4-2 I

SONGS 2 HEATUP RCS PRESSURE / TEMPERATURE

^

LIMITATIONS FOR S UNTIL:20 EFPY l Guff Normal Operation i SAN ONOFRE-UNIT 2 3/4 4-29 AMENDMENT NO.

1 l

l

_ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ J

(Figure 3.4 Not Used) l L

SAN ONOFRE-UNIT'2 l3/4:4229a1 AMENDMENTE N0;

l l

l 3500 , ,

LOWEST SERVICE COOLDOWN TEMP = 209 F i

3000 - 8 -

t i

I

,> 4 9 ., , 8..

e g 2500 --

l 8 --

a i

D i (n i Ch 3 2000 l 0- ^

Unacceptable '

@ Operating t2 Region t '

D Acceptable (h 1500 - - > - -

r Operating -

$ l Region 6% I C a O. "e

.s a . . . ,

L.O g - 's -

< 1000 .. -

9 I O t E

500 -

f -

MINIMUM i + *- ++' -* * -- '-- --

BOLTUP i s TEMP = 86*F _

h ,

. . ..Ii.i iIii;,Iiii I;i;iI,,,.Ii;, 1,,,,

0 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE (*F)-Tc FIGURE 3,4--3 3.4-4 SONGS 2 000LDOWN'RCS PRESSURE / TEMPERATURE LIMITATIONS'F0" ^ S UNTil 20 EFPY l$pfi Normal Operation g SAN ONOFRE-UNIT 2 3/4 4-30 AMENDMENT NO.

4l 120 ,i .i -

i >s .i.i. - g -

i - , .

110 100 no _

E m - -

I it

$ m - -

4 -

g 80 - -

g .

o g _ _

S .

S4fl 8o 40 - -

30 . _

20 -

10 .

o i .i.i.i . i.i i i i . i . i i .

l 80 90 100 110 120 130 140 150 160 170 180- 190 200 210 INDICATED RCS TEMPERATURE ('F)-Tc NOTE:2A' MAXIMUM.C00LDOWN RATE 0FJ100*F/HR:IS' ALLOWED l ATANYTEMPERATURE$80VEjl60*F .)6pf)

FIGUREe3 J4-5 SONGS'2 RCS: PRESSURE / TEMPERATURE:': LIMITS

MAXIMUMALLOWABLEC00LDOWN: RATES 1(UNTIU20.EFPY)

[stp l

~NormahOperation~"

-SAN'ONOFRE-UNITJ2 L3(4'4230a : AMENDMENT NO.-

l zM , ,

,,,,...i,,,,i,,,,i,,,,i,,,,i,,,,i,,,,

LOWEST SERVICE COOLDOWN TEMP 209 F 4

4 3000 .. .o a. . ......... . . . ..u . a..

t' l

s m

_.i.. . .

Q- 2500 -- 8, w a.

I C '

D W .,

C a 2000 -

' Unacceptable

~~

~

@ l Operating

!S Region C e D i Acceptable 15M - - ,

e i Operating Q-l Region o i W i H i 4 4 1000 . .

3 1 Soo . _.. * . . . .

MINIMUM BOLTUP .; .. .._ . .

TEMP = 86*F A

,3

. . i.I.i. .t... 1. ...t....t . . . . i....i....

0 50 100 150 200 250 300 350 400 IND!CATED RCS TEMPERATURE ( F)-Tc FIGURE 3.4-6 SONGS 2 C00LDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY Remote Shutdown Operation UTg SAN ON0FRE-UNIT 2 3/4 4-30b AMENDMENT NO.

i l

__. - - - _ _ ____ _ -__-_______- - -________ - ___- - -_ - - _ .- -__-____- _ ____ a

I

-i

= .

l 110 100 -

90 - -

80 -

70 -

T ~

k -

w 60 -

LLI

$ll? !

<C 50 x "

Z 3 40 - -

o 3 '

O 30 - -

O O -

20 -

10 O I I

I I I i > l 8 I I > I - I -

80 90 100 110 120 130 140 150 160 170 180 190 200 210 INDICATED RCS TEMPERATURE ( F)-Tc

, NOTE::A~MAXIMUN'C00LDOWN RATE'~0F0100*F/HR ISTALLOWED l AT:ANYTEMPERATURE/AB0VE168*F lGotf l FIGU3El3.4-7

$0NGS'2"RCS' PRESSURE / TEMPERATURE? LIMITS MAXIMUM ALLOWABLCC00LDOWN RATES;;(UNTIL:l20: EFPY)~ pgp g-Remote (Shutdown;0peration' SAN ONOFRE-UNITL2' 3 3/4 l4-30c:- _

JAMENDMENT NO.

=

TABLE 3.4-3 Low Temperature RCS Overoressure Protection Range Operating Period. EFPY Cold Leo Temperature, oF During During Heatup Cooldown 4-t+40 Until 20 (NormalT0peration) s ale 256 s B&F 238 SufI Until 20 (Remote Shutdown Operation)- '*

's 238

~

Heatup operations are not'normally performed from-the' Remote Shutdown panels.

SAN ONOFRE-UNIT 2 3/4 4-30ad AMENDMENT NO.

'1 ' _- _ __ -_________ .

s REACTOR C00LAN1 SYSTEM OVfRPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s 317 256of 9/f LIMITING CONDITION FOR OPfRATION 3.4.8.3.1 No more than two high-pressure safety injection pumps shall be OPERABLE and at least one of the following overpressure protection systems shall be OPERABLE:

a. The Shutdown Cooling System Relief Valve (PSV9349) with:

1) A lift setting of 406 1 10 psig', and

2) Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377, and _

2HV9378 open or,

b. The Reactor Coolant System depressurized with an RCS vent of greater than or equal to 5.6 square inches.

APPLICABILITY: MODE 4 when the temperature of any one RCS cold leg is less than or equal to the enable temperatures specified in Table 3.4-3; MODE 5; and MODE 6 when the head is on the reactor vessel and the RCS is not vented.

ACTION:

to less than 200of,

a. With the SDCS depressurize andRelief Valve vent the RCSinoperable, through a greater reducet T,,han or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one or both SDCS Relief Valv> isolation valves in a single SDCS Relief Valve isolation valve pf (valve pair 2HV9337 and 2HV9339 or valve pair 2HV9377 and 2HV9378) tosed, open the closed valve (s) or power-lock open the OPERABLE SDCS Relief Valve isolation valve pair within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or reduce T oy to less than 200af, depressurize and vent the RCS through a greater than or equal to 5.6 inch vent within ;

the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. With more than two high-pressure safety injection pumps OPERABLE, secure the third high-pressure safety injection pump by racking out its motor circuit breaker or locking close its discharge valve within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The 'lif t . setting pressure applicable to-For valve temperatures of less than or equal to 130oF.

SAN ONOFRE-UNIT 2 3/4 4-32 AMENDMENT NO.

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE >3 M 256af lgng /

tIMITING CONDITION FOR OPERATION 3.4.8.3.2 At least one of the following overpressure protection systems shall be OPERABLE:

a. The Shutdown Cooling System Relief Valve (PSV9349) with:

1) A lif t setting of 406 2 10 psig*, and

2) Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377, and 2HV9378 openr --err or, yl

b. A minimum of one p,ressurizer code safety valve with a lif t setting of 2500 psia i 1%

APPLICABillTY: MODE 4 with RCS temperature above that specified in Table l 3.4-3.

ACTION:

a. With no safety or relief valve OPERABLE, be in COLD SHUTDOWN and vent the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

b. In the event the SDCS Relief Valve or an RCS ved is used to mitigate an l RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the SDCS Relief Valve code safety valve ENC-S-vent on the I transient and any corrective action necessary to prevent recurrence.

, SURVEILLANCE REQUIREMENTS 4.4.8.3.2.1 The SDCS Relief Valve shall be demonstrated OPERABLE by:

a. Verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that the SDCS Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 are open when the SDCS Relief Valve is being used for overpressure protection.

b. Verifying relief valve setpoint at least once per 30 months when tested pursuant to Specification 4.0.5 4.4.E.3.2.2 The pressurizer code safety valve has no additional surveillance requirements other than those required by Specification 4.0.5.

4r4.8.3.2.3 The-RGS-vent-sha44-be ver i F4ed-to-be-open-at4 east-onee-pw-M-hours when-t he-vent 4+-beite-used-fee-overpressure-proteet4eweept-when-the-vest paOway4+-prended-Mth-a-vake-wMeh-4s-leeked, scaledree-ethetw+se-seeeeed4n

.the--open-pn94-t4eny-then-ver4fy-these-vahes-open at least onec pcr 31 daysv

" The. lift setting pressure applicable to-Fee valve temperatures of less than or equal to 130of.

"'The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SAN ONOFRE UNIT 2 3/4 4-33 AMENDMENT NO.

REACTOR COOLANT SYSTEM I

BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The heatup and cooldown limit curves for_~normalfoperation (Figures 3.4-2 and L4-3;3.4 4) and-the cooldowntlimit"curselfor- remote shutdown operation ((Figure 3.4-6)"are composite curves which were" prepared'by determining th'eimost >

conservative case, with either the inside or outside wall controlling, for any heatup rate of up to 60aF/hr or cooldown rate of up to 100aF/hr. The limit curves for Remote Shutdowntoperationf are determined 1using theiTotal Loop Uncertainties-(TLUs)lfor temperature'and pressure for the Remote Shutdown 1 Panel instruments in-which;the-pressure.TLUs are higher than those for the Control Room shutdown-instruments. The heatup and cooldown curves were prepared based'upon the' most' limiting' value of the predicted adjusted reference temperature at the end of the service period, and they include adjustments for poss4Me-er-rer+-4n the-pre %ure-and temper *teve-sens4ng-inst +ument+ instrumenti uncertainties', and ~

=_

static'and dynamic heads.

The reactor to determine vessel their initial materials RTgo7; the resulhave-been L'of these tests were tested and;the prio'r;toFreactor updates ~~ start resulting fromithe(evaluation of material properties)in response to; Generic Letter 92-01',;" Reactor ; Vessel; Structural Integrity," -Revision' l'are shown in Table B 3/4.4-1; Reactor '~ operation and resultant fast neutron"(E greater than 4 1: MeV) irradiation will cause an increase in the RTsar. Therefore, an adjusted reference temperature, based upon the fluence and copper and nickel content of the material in question, can be predicted using FSAR Table 5.2-5 and the recommendations of Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials." The heatup limit curvet (FigureL3 4-2) and the cooldown limit curves, Figures 3.4-3 4?an_d 3.4-6, include predicted adjustments for this shift in RT at the end of the applicable service period, as well as adjustments for pos dge-erroes-in-the pressure and temperatere sensing inst +uments instrument! uncertainties, andl static and; dynamic' heads.

The actual shift in RTnar of the vessel material will be established periodically during operation by removing and evaluating, in accordance.with ASTM E185-73 and 10 CFR 50 Appendix H, reactor vessel material irradiation sur-veillance specimens installed near the inside-wall of the reactor vessel in the core area. The surveillance specimen withdrawal schedule is shown in Table 4.4-5. Since the neutron spectra at' the-irradiation samples and vessel inside radius are essentially identical, the measured transition shif t for a-sample can be applied with confidence to the adjacent section of the reactor vessel taking into account the location of the sample closer to the core than the vessel wall-by means of the Lead Factor. The heatup and cooldown curves so mustbe'recalculatedwhenthedeltaRT{determinedfromthesurveillance capsule is different from the calculatec delta RTuor for the equivalent capsule [-

radiation exposure.

The pressure-temperature limit lines shown on Figure 3.4-2 and 3.4-3 fo'r reactor criticality and for inservice leak and hydrostatic testing have been l provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

SAN ON0FRE-UNIT 2 B 3/4 4-7 AtlENDMENT NO.

l REACTOR C00lANT SYSTQ1 1

RASFS PRESSURE / TEMPERATURE LIMITS (Continued)

The maximum RT,,31 for all reactor coolant system pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 900F. The Lowest Service Temperature limit line shown on Figures 3.4-2, 3.4-4, and 3.4-3 6 is based upon this RT e1 since Article NB-2332 (Summer Addenda yl of 1972) of Section 111 of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTyr + 1000F for piping, pumps and valves.

Below this temperature, the system pressure must be limited to a maximum of 20%

of the system's hydrostatic test pressure of 3125 psia.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature dif ferential are provided to assure that the pres-surizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The Low Temperature Overpressure Protection (LTOP) enable temperatures are based upon the recommendations of NUREG-0800 Branch Technical Position (BTP) RSB 5-2, Revision 1, "0verpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures." BTP RSB 5-2, Revision 1 defines the enable temperature as "the water temperature corresponding to a metal temperature of.at least RT y + 90af at the beltline location (1/4t or 3/4t) that is controlling _ in theAppendixGlimitcalculations."

SAN ONOFRE-UNIT 2 B 3/4 4-7a AMENDMENT NO.

l TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS

Temperature of Minimum Upper )

Drop Charpy V-Notch Shelf Cv energy '

@ Weight 0 30 0 50 for Longitudinal

? Results Direction-ft Ib i Piece No. Code No. Material Vessel Location ft - lb - ft - lb

=

Upper Shell Plate 40 15 35 130 215-01 C-6403-1 A533GRBCL-1 Upper Shell' Plate 0 20 25 133 215-01 C-6403-2 A533GRBCL-1

" Upper.Shell: Plate -10 20 45 131 215-01 C-6403-3 A533GRBCL 1 Intermed. Shell Plate -30 M 40 50 80 445 119 215-03 C-6404-1 A533GRBCl-1 215-03 C-6404-2 A533GRBCL-1 Intermed. Shell! Plate -20

-20 M 70 M 70 50 80 50 80 M 5 113 B4 99

[9f'l 215-03 C-6404-3 A533GRBCL-1 Intermed. Shell Plate Lower Shell Plate -10 --5 -40 25 80 M4 104 215-02 C-6404-4 A533GRBCL-1 B4 118 C-6404-5 A533GRBCL-1 Lower Shell Plate -20 M '50 35 70 215-02 454 124 215-02 C-6404-6 A533GRBCL-1 Lower Shell' Plate

-10 -20 50 0 50 C-6401 A508Cl-2 Vessel Flange Forging -10 -70 -35 148

[ 238-02 s

" A508Cl-2 Closure Head Flange -10 -90 -40 142 l 209-02 C-6402 i Forging

[

A508Cl;2 Inlet Nozzle Forging 20 -40 -35 130 205-02 C-6410-1 Inlet Nozzle Forging 0 -20 -5 135 205-02 C-6410-2 .A508Cl-2 C-6410-3 A508Cl-2 Inlet Nozzle Forging 0 -15 -15 140 205-02 -50 140 C-6410-4 A508Cl-2 Inlet Nozzle Forging 0 -65 205-02 Outlet Nozzle Forging -100 -30 -10 140 205-06 C-6411-1 A508C112

-10 140 205-06 C26411-2 A508CI-2 Outlet _ Nozzle1 Forging 0 -35 lG{l Bottom Head Torus -50 -20 10 122 232-01 C-6424 A533GRBCL-1

-50 -30 -20 136

> 232-02 C-6425 A533GRBCL-1 Bottom Head Dome l 5 Inlet Nozzle Forging S/E -30 -70 -50 174 5 205-03 C-6428-1 _A508CL-1 Inlet Nozzle Forging lS/E -30 -70 50 174 C-6428-2 A508Cl21

$ 205-03 Inlet:Nozzlefforging $/E -30 -70 -50 174 Z 205-03 C-6428-3 A508CL-1 C-6428-4 A508Cl-1 I_let'NozzleLForginglS/E n -30 -70 -50 174 g 205-03

__u

UMWMMMM l

TABLE B 3/4:4-l (Continued)

I I

Temperature of Minimum Upper

$ Drop Charpy V-Notch Shelf Cv energy

] Weight 0 30 0 50 for Longitudinal

=

Vessel Location Results ft - lb - ft - lb Direction-ft lb .giff

, y Piece No. Code No. Material Im Outlet Nozzle Ext. Forging -30 -40 -25 229 I l E- 205-07 C-6429-1 A508CL-1 Outlet Nozde Ext. Forging -30 -40 -25 229 205-07 C-6429-1 A508CL-1 l l*

l " +10 20 55 118 231-02 C-6430-1 A533GRBCL-1 Closure Head Peels

-20 10 50 100 231-02 C-6431-1 A533GRBCL-1 Closure Head Peels

-10 -15 45 115 231-02 C-6432-1 A533GRBCL-1 Closure ' Head . Peel s

-10 -15 45 115 231-02 C-6432 A533GRBCL-1 Closure Head Dome

.cm LJ

?

2' 35 9

a I

8

ENCLOSURE 3 TECHNICAL SPECIFICATION PAGES CONTAINING THE CHANGES WHICH WERE PREVIOUSLY REQUESTED IN AMENDMENT APPLICATION NO. 117 (PCN-354) DATED SEPTEMBER 3, 1992, $uf l AND ARE BEING REQUESTED IN THIS LICENSE AMENDMENT APPLICATION NO. 118 (PCN-335)

UNIT 2 l$p l

1 INDEX LIST OF TABLES TABLE PAGE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION....................... 3/4 3-52 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................. 3/4 3-54 3.3-11 FIRE DETECTION INSTRUMENTS-MINIMUM-MSTRUMENTS OPERABL-E. . . 3/4 3-57 3.3 le RADIGAC41VE-t-lQMD-EFFEUENT-MONMORTNG-INSMUMEMTAT40N-l -oEmsa- h0g 4r3 8 RADIOAGT4#E-4-IQ410 EFFLUENT-NONITORING INSTRUt4ENTAT40N

---SURVE414-ANGE-REQUMENENTS DELETED 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION... 3/4 3-65 l 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................. 3/4 3-67 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION...................................... 3/4 4-14 4.4-2 STEAM GENERATOR TUBE INSPECTION........................... 3/4 4-15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.......... 3/4 4-19 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY.......................... 3/4 4-21 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE......... 3/4 4-30ad [

4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS.............................................. 3/4 4-22 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM................................................... 3/4 4-25 4,4 5 REAC40R-VE&SEL-MATEMAL-SURVEILLANCE PROGRAM W 14 HDR AWAl-SC44EDU L E . . . . . . . . . . . . . . . ., . . . . . . . . . . . . . . . . . . . . . . 3/4 ^ 28 Y 4.6-1 TENDON SURVEILLANCE....................................... 3/4 6-12 4.6-2 TENDON LIFT-0FF F0RCE..................................... 3/4 6-12a 3.6-1 CONTAINMENT IS0tATION VALVES.............................. 3/4 6-20 3.7-1 MAIN STEAM SAFETY VALVES.................................. 3/4 7-2 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS................................ 3/4 7-3 SAN ON0FRE-UNIT 2 XIX AMENDMENT NO.

b _.

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS ;

REACTOR COOLANT SYSTEM l

SURVElLl ANCE REQUIREMENTS Mont4nued)

'GUfI 4.4.8.1.1 The Reactor Coolant-System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, -to determine changes in material propertiesr-at-the 4ntervalsias required by 10 CFR 50 Appendix H. ir accordance seith the schedule in Tame-4r4-6 The results of- these examinations shall be used to update Figures 3.4-2 and 3r4-3f 3.4-4' through ~314-7. Recalculate the' Adjusted Reference- ,

Temperature based-en-the-greater 'cf the following:JinTaccordance;with: Regulator 9 ~

Guide 1.99, : Revision 2; "RadiationiEmbrittlementrof. Reactor-Vessel-:1 Materials," May og_ -

i 1988.- I' I a,--The mean salue of -hif t in ref erence temperature for plate -

C-64M-3*r-or -

b. The predicted shif t in reference temperature for iJeld scam: 3 203A, OF 3-2MB as determined bj Rcwlatcry Guide 1.99, Revision 2, "Radiaticm-Embrittlement of Reastcr-Vessel Materials," May 1988.

I i

l 1

3hc ment-limiting mater 4al in the reactor vessel in accordance ..ith the new Regulatory Guide 1.99, Revisien-G, " Radiation Embrittlement of Reacter Vessel Materials," May -1988, has-ehanged-and-are- plate C 5101 3. Cal cul a ti ve . p rocc--

dares-prCVided in-the nc% guide should bc used te cblain the 00an values of- G )

shift in Rim -ef C 5101 3-phten. Calculation are based en the actual shift- l 4n-referenee-temperatere as determined by impact test 49-er, the existing plate ~

C-6404 2 ;urve444ance-eaterial .

- SAN ONOFRE-UNIT 2 3/4 4-27a 128- AMENDMENT NO.

M l

l l

_ _ _ _ . _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ d

m

~c.

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The heatup and cooldown limit curves for ' normal operation (Figures 3.4-2 and pcN 3A--3 3.4-4) and the cooldown limit curve for remote shutdown operation (Figure gf 3.4-6) are composite curves'wh'ich were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate of up to 600F/hr or cooldown rate of up to 100cF/hr. The limit-curves.for. Remote Shutdown operation are determined-using the Total. Loop 17M Uncertainties-(TLUs)~for temperature and pressure for the Remote Shutdown Panel l' /

instruments in which the pressure TLus are higher than those for the_ Control-Room

~

N shutdown-instruments. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period, and they include adjustments for possible cr-res4n- L the-pressure-and-temperatur-e-sens4ng-4nstement-s instrument uncertainties, and Cd static and dynamic heads. f.gf The reactor vessel materials hwe-been were tested prior to reactor startup c to determine their initial RT s the results of these tests and the updates lh,5' g resulting from the' evaluation m;of material! properties in response to Generic Letter 92-01, " Reactor Vessel Structural-Integrity," Revision 1 are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (L greater than d' g 1 MeV) irradiation will cause an increase in the RT sm. Therefore, an adjusted reference temperature, based upon the fluence and copper and nickel content of the material in question, can be predicted using FSAR Table 5.2-5 and the recommendations of Regulatory Guide 1.99, Revision 2 " Radiation Embrittlement of Reactor Vessel Material 3." The heatup limit curve (Figure 3.4-2) and the cooldown limit curves, figures 3.4-3~4 and 3.4-6, include predicted adjustments lvci[ IlW for this shift in RT 3 at the end of the applicable service period, as well as adjustments for powdle-eFren-4n-the-prewur+end-temperature sensing 4nstw eents instrument uncertainties, and static and dynamic heads.

h 3M The actual shift in RT ug of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR 50 Appendix H, reactor vessel material irradiation sur-veillance specimens installed near the inside wall of the reactor vessel in a the core area. The surveillance specimen withdrawal schedule is shou in f0V Tabic 4.4 5. maintained in.the FSAR. Since the neutron spectra at the irradiation pf samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel taking into account the location of the sample closer to the core than the vessel wall by means of the Lead Factor. The heatup and cooldown curves must be recalculated when the delta RT s3 determined from g the surveillance capsule is different from the calculated delta RT equivalent capsule radiation exposure.

3g for the h

The pressure-temperature limit lines shown on Figure 3.4-2 and 3.13 for reactor criticality and for inservice leak and hydrostatic testing have been lfN provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

SAN ONOFRE-UNIT 2 B 3/4 4-7 AMENDMENT NO.

1

_ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . .

ML20071F822

Text

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